This invention relates to anaylsis of fuel rod performance in a nuclear power plant reactor and more particularly to the monitoring of power density in a nuclear power reactor for analysis of fuel rod performance.
Fuel rod failure is a costly factor in nuclear power plant operation and has created a need for the development of an on-line, fuel failure avoidance system through which (a) the state of all fuel rods is continuously monitored, (b) the data obtained by monitoring is analyzed, and (c) fuel rod failure forecasts are generated as a result of such analysis.
The foregoing anaylsis function of such failure avoidance systems has involved the creation of a failure model from which to calculate the expected frequency of fuel failure in commercial nuclear power plant reactors. One such failure model developed is based on the concept of fuel rod failure resulting primarily from pellet-clad interaction. Experiments have shown that pellet-clad interaction failures occur either during rapid increases in power, referred to as "power shocks", or within a few hours thereafter. The power shocks produce thermal expansion, fission gas release, and shape distortion of the fuel pellets causing clad strain and longitudinal cracks in the fuel rod cladding. Pellet deformation also results in axial localization of strain at pellet joints. Other factors associated with pellet clad interaction are also believed to be responsible for fuel rod failure, but all such factors are consequences of power shocks.
Accurate and continuous monitoring of fuel rod power density is therefore essential in order to enable detection of power shocks and through a failure model as aforementioned to furnish a power utility operator with the information necessary to control power distribution by control rod movement and/or coolant flow rate control in a boiling water reactor or by control rod movement and/or boron concentration control in the moderator of a pressure water reactor.
The power density of the reactor has been monitored through sensors located in the fuel rod assembly. One type of sensor heretofore utilized for such purpose has been of the thermal neutron flux type. Although such neutron flux sensors provide power measurement signals that exhibit a rapid response to changes in local power density, they are unsatisfactory from two other important standpoints. First, the power measurement signal of a neutron flux sensor is not directly related to the linear heat generation rate of the fuel rod so that various calibration and correction factors must be introduced in order to approximate the rather complex relationship involved. Second, the neutron flux sensor has an emitter subject to burn-out.
According to prior copending application Ser. No. 888,881, filed Mar. 21, 1978 now U.S. Pat. No. 4,298,430, issued Nov. 3, 1981 owned in common with the present application, a local power density sensor is disclosed, which provides a signal output which is directly related to the linear heat generation rate for the fuel rods to enable more accurate determination of this parameter as compared to measurement by neutron flux sensors. Further, the sensor disclosed in the aforementioned prior patent is of the gamma radiation heat generating type which has no emitter subject to burn-out. However, the gamma ray sensor does not have a rapid signal response to changes in power as in the case of a neutron flux sensor which is in conflict with the requirement for real time local fuel power measurements in a fuel failure avoidance system.
It is therefore an important object of the present invention to provide a method of furnishing a nuclear power utility operator with real time, yet accurate, knowledge of local fuel power rate in the reactor core to enable operation of the power plant within adequate margins with respect to those operational parameters determined from local power measurements.
An additional object is to provide a power monitoring system for the fuel rods of the nuclear power reactor which benefits from the use of a gamma ray type of sensor.